What I have learned from EAST experiments 胡友俊

Slides:



Advertisements
Similar presentations
Korean Modeling Effort : C2 Code J.M. Park NFRC/ORNL In collaboration with Sun Hee Kim, Ki Min Kim, Hyun-Sun Han, Sang Hee Hong Seoul National University.
Advertisements

Enabling Technology on Alcator C-Mod Jim Irby MIT-PSFC.
XP 1157 Increasing the CHI start-up current magnitude in NSTX B.A. Nelson et al. 1.
Disruption Analysis of PP, VV, and Components Model includes bracket conducting path on back of PPs Disruption cases: Mid plane fast disruption (1 ms)
Critical Issue on Plasma Equilibrium for CS-less Tokamaks Kichiro Shinya Toshiba Corporation Japan-US Workshop on Fusion Power Plants and Related Advanced.
Physics of fusion power Lecture 14: Anomalous transport / ITER.
Physics of fusion power
Physics of fusion power
Magnet System Definition L. Bromberg P. Titus MIT Plasma Science and Fusion Center ARIES meeting November 4-5, 2004.
Development of the New ARIES Tokamak Systems Code Zoran Dragojlovic, Rene Raffray, Farrokh Najmabadi, Charles Kessel, Lester Waganer US-Japan Workshop.
Physics of fusion power Lecture 8 : The tokamak continued.
Physics of fusion power Lecture 10 : Running a discharge / diagnostics.
Physics of Fusion power Lecture 7: Stellarator / Tokamak.
Magnetic Diagnostics for GLAST-III Tokamak M. A. Naveed, Aqib javeed and GLAST Team National Tokamak Fusion Programme Islamabad Pakistan IAEA First Technical.
托卡马克的平衡计算 李国强 四室学术报告. Introduction Decompose the physics problem by the orders (time order and space order) Traditional decomposition of plasma.
ASIPP EAST Overview Of The EAST In Vessel Components Upgraded Presented by Damao Yao.
NSTX_U Design Point Study Active H2O Cooling Pulse Length 60 sec C Neumeyer 5/19/6.
1 Modeling of EAST Divertor S. Zhu Institute of Plasma Physics, Chinese Academy of Sciences.
V.I. Vasiliev, Yu.A. Kostsov, K.M. Lobanov, L.P. Makarova, A.B. Mineev, D.V.Efremov Scientific Research Institute of Electrophysical Apparatus, St.-Petersburg,
1 CHI Summary Transient CHI (XP606) –All systems operated reliably without any faults Edge Current drive (XP533)
Remote operation of the GOLEM tokamak Gent University, December 14, 2009 Jan Stockel Institute of Plasma Physics, Prague Vojta Svoboda.
ARIES AT Project Meeting - Princeton, NJ 18 Sept 00 1 ARIES-AT Toroidal Field (TF) and Poloidal Field (PF) Coils Tom Brown, Fred Dahlgren, Phil Heitzenroeder.
US ITER UFA Meeting APS-DPP Savannah, GA Ned Sauthoff (presented by Dale Meade) November 15, 2004 US In-kind Contributions and Starting Burning Plasma.
12/03/2013, Praga 1 Plasma MHD Activity Observations via Magnetics Diagnostics: Magnetic island Analysis Magnetic island Analysis Frederik Ostyn (UGent)
Physics of fusion power Lecture 10: tokamak – continued.
V. A. Soukhanovskii 1 Acknowledgement s: R. Maingi 2, D. A. Gates 3, J. Menard 3, R. Raman 4, R. E. Bell 3, C. E. Bush 2, R. Kaita 3, H. W. Kugel 3, B.
Divertor Design Considerations for CFETR
OPERATIONAL SCENARIO of KTM Dokuka V.N., Khayrutdinov R.R. TRINITI, Russia O u t l i n e Goal of the work The DINA code capabilities Formulation of the.
PF1A upgrade physics review Presented by D. A. Gates With input from J.E. Menard and C.E. Kessel 10/27/04.
Physics of fusion power Lecture 9 : The tokamak continued.
EAST Data processing of divertor probes on EAST Jun Wang, Jiafeng Chang, Guosheng Xu, Wei Zhang, Tingfeng Ming, Siye Ding Institute of Plasma Physics,
HT-7U CASIPP 1 The Engineering Design of the Poloidal Field System of HT-7U Tokamak PF Group of HT-7U Team Chinese Academy of Sciences, Institute of Plasma.
ASIPP Long pulse and high power LHCD plasmas on HT-7 Xu Qiang.
CHI Run Summary for March 10-12, 31 & April 9, 2008 Flux savings from inductive drive of a Transient CHI started plasma (XP817) R. Raman, B.A. Nelson,
OPERATIONAL SCENARIO of KTM Dokuka V.N., Khayrutdinov R.R. TRINITI, Russia O u t l i n e Goal of the work The DINA code capabilities Formulation of the.
1) Disruption heat loading 2) Progress on time-dependent modeling C. Kessel, PPPL ARIES Project Meeting, Bethesda, MD, 4/4/2011.
Design Point Studies for next step device National High-heat-flux Advanced Torus Experiment NHTX C Neumeyer 6/8/6.
Efremov Research Institute Russia, St. Petersburg, FAX: (812) ,
Comprehensive ITER Approach to Burn L. P. Ku, S. Jardin, C. Kessel, D. McCune Princeton Plasma Physics Laboratory SWIM Project Meeting Oct , 2007.
Rajesh Maingi Oak Ridge National Laboratory M. Bell b, T. Biewer b, C.S. Chang g, R. Maqueda c, R. Bell b, C. Bush a, D. Gates b, S. Kaye b, H. Kugel b,
Work with TSC Yong Guo. Introduction Non-inductive current for NSTX TSC model for EAST Simulation for EAST experiment Voltage second consumption for different.
Charles C. Baker Deputy US ITER Planning Officer presented at the Fusion Power Associates Annual Meeting and Symposium Washington, DC November, 2003.
08/07/05Plasma Operations : Introduction P.J.Lomas 1 Plasma Operations: General Introduction Peter Lomas and Plasma Operations Group:- Jerzy Brzozowski,
1 EAST Recent Progress on Long Pulse Divertor Operation in EAST H.Y. Guo, J. Li, G.-N. Luo Z.W. Wu, X. Gao, S. Zhu and the EAST Team 19 th PSI Conference.
GOLEM operation based on some results from CASTOR
Instituto Nacional de Pesquisas Espaciais – Laboratório Associado de Plasma 1 3rd IAEA TM on ST – STW2005 Determination of eddy currents in the vacuum.
ASIPP Magnetic Diagnostics of HT-7U Tokamak Shen Biao Wan Baonian Institute of Plasma Physics, CAS P.O.Box 1126, Hefei, Anhui , P.R.China (e_mail:
Solenoid Free Plasma Start-up Mid-Run Summary (FY 2008) R. Raman and D. Mueller Univ. of Wash. / PPPL 16 April 2008, PPPL 1 Supported by Office of Science.
1 PFC requirements  Basic requirements  Carbon based  Provisions for adding (interface design included in research prep budget)  NBI armor  Trim coil.
Plasma MHD Activity Observations via Magnetic Diagnostics Magnetic islands, statistical methods, magnetic diagnostics, tokamak operation.
Open house 2013 slides T. Brown. 1.6-m FNSF Super-X Divertor Configuration.
Simulation of Non-Solenoidal Current Rampup in NSTX C. E. Kessel and NSTX Team Princeton Plasma Physics Laboratory APS-DPP Annual Meeting, Savannah, Georgia,
Design Point Studies for Next Step Device National High-power Advanced Torus Experiment NHTX C Neumeyer 7/25/06.
Current Drive Experiments with Oscillating Toroidal Flux in HT-7 Superconducting Tokamak J.S.Mao, P. Phillips 1, J.R.Luo, J.Y.Zhao, Q.L.Wu, Z.W.Wu, J.G.Li,
Long Pulse High Performance Plasma Scenario Development for NSTX C. Kessel and S. Kaye - providing TRANSP runs of specific discharges S.
Status of Upgrading Project of Tokamak T-15 E. Azizov 1), P. Khvostenko 1), I. Anashkin 1), V. Belyakov 2), E. Bondarchuk 2), O. Filatov 2), V. Krylov.
Disruption Analysis of PP, VV, and Components. Opera 3D Model – Transient ELEKTRA Solver Fast mid-plane centered disruption 2 MA/ms Back ground field.
16 th IEEE NPSS Real Time Conference 2009 May 10 – 15, 2009 IHEP Beijing China ASIPP Current Status of EAST Plasma Control and Data Acquisition Bingjia.
Enhancement of EAST plasma control capabilities
Instrumentation for status monitoring of SST-1 superconducting magnets
MANUFACTURING OF MAGNETS FOR SST-1 TOKAMAK
JongGab Jo, H. Y. Lee, Y. H. An, K. J. Chung and Y. S. Hwang*
Construction and Status of Versatile Experiment Spherical Torus at SNU
P.Sonato On behalf of RFX-Team
Control and data acquisition system of the KTX device
Design and Fabrication of Versatile Experiment Spherical Torus (VEST)
X.R. Wang, M. S. Tillack, S. Malang, F. Najmabadi and the ARIES Team
Lecture 09 - Inductors and Capacitors
Progress on Systems Code Application to CS Reactors
Valves according to ISO Standard
Presentation transcript:

What I have learned from EAST experiments 胡友俊 2016-04-01

Main Coils on Tokamak A. Boozer, 2004, Rev. Mod. Phys.

Toroidal Field (TF) coils of EAST TF coils=16coil*130turn/coil Two scenarios for EAST: =10000A = 2.447T at R=1.7m = 8000A  = 1.957T at R=1.7m

Poloidal Field (PF) Coils of EAST 12 superconductive PF coils (6 central solenoid coils+6 shaping coils) PF1 to PF6120 turns/coil PF7+PF9 connected in series248 turns PF8+PF10 connected in series248 turns PF1164 turns PF1264 turns PF1332 turns Pf1432 turns Maximal current per turn=14.5kA 2 in-vessel Copper coils (2 turns/coil) connected in anti-series to control VDE

Plasma Control System (PCS) on EAST Adapted from DIII-D PCS Acuators: 12 PF coils + 2 fast coils Two typical control scenarios: “RZIP control” (R,Z) of the center of the plasma current, and the total current IP are feedback controlled by PF coils. “iso-flux control” shape of LCFS, in additon to the RZIP, is feedback controlled by PF coils.X point, gap between plasma and wall

Typical wave-forms of current in PF coils Charge to maximal current before plasma discharge to provide maximal Voltage*Second Central solenoid coils Charge to median current before plasma discharge to provide flexibility of shaping plasma Plasma Discharge beginning Shaping coils

Components and material of first wall of EAST divertor plate (W) High-field-side wall (Mo+Carbon) Ports for diagnostic and heating (antenna, NBI) Inner divertor plate (C) Passive plates (Mo) Outer divertor plate (Carbon) Dome (Carbon)

Time evolution of space-averaged quantities Complete signal names on MDSplus server can be found on EAST wiki page View data online: webscope Download data for analysis: via webscope directly access MDSplus server using matlab or Fortran code

View Plasma configuration: “eastviewer” on cs1.ipp.ac.cn

View profile: EASTprofiles on cs1.ipp.ac.cn

谢谢!

Gass puff, LHW asistent breakdown and start-up